1. Field of the Invention
The invention generally relates to methods and systems for the generation of radioisotope products and in one embodiment to the generation of technetium-99m.
2. Background Information
Technetium-99m (Tc-99m), the radioisotope most widely used in nuclear medicine diagnostic procedures, is used in the detection of cancer, heart disease and thyroid disease, along with the study of brain and kidney function and the imaging of stress fractures. Over the past two decades, the nuclear medicine industry has experienced intermittent shortages of molybdenum-99 (Mo-99), the parent of Tc-99m as a result of unplanned reactor outages. The world's supply of medical Mo-99 is primarily produced by five aging reactors in Canada, Europe, and South Africa, and at present no strategy exists to provide a global, long-term reliable supply of Mo-99. In addition, in the interest of nuclear security and non-proliferation, the U.S. and other countries are increasing the pressure to migrate the industry from using customary Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) for the production of Mo-99.
There are a variety of available Mo99 production methods including U235(n,f)Mo99, U238(γ,f)Mo99, Mo98(n,γ)Mo99, and Mo100(γ,n)Mo99. A National Academy of Sciences study commissioned in 2009 recommended the irradiation of LEU in a fission reactor to produce Mo-99 where separation of Mo from the plurality of uranium fission products and actinides continues to be required to obtain high specific activity Mo-99. This makes the “fission product” Mo method economically challenging in view of the dedicated specialized radiochemical facilities and equipment and the increasing regulatory demands associated with uranium target manufacturing and the handling, storing, and disposing of the nuclear waste from fissioned uranium. There is a need for a more economical method to produce Tc-99m from Mo-99.
Producing Mo99 from natural or isotope-enriched Mo targets has the main advantage of eliminating the need for dedicated radiochemical facilities (hot cells) while producing nearly no waste stream compared with the fission product approach. For example, Mo99 produced from neutron capture of natural Mo (containing about 24% Mo98) provides certain advantages and is discussed herein in relationship to the invention (it is understood that the invention is not limited to such production mode of Mo99). There are two disadvantages to using Mo99 derived from neutron capture of natural molybdenum for Tc-99m generation. First, the low specific activity of such “neutron-capture” Mo-99 (i.e., number of curies of Mo99 per gram of Mo) typically requires very large Tc-99m generators and elution volumes when compared to conventional alumina generators using fission product Mo-99. Second, the separation of Tc-99m from the parent Mo-99 in a compact generator requires high selectivity (USP guidelines indicate that there must be less than 0.15 microcuries of Mo99 per millicurie of Tc-99m and no more than 10 micrograms of aluminum ion per milliliter of generator eluate). No effective ion exchange technology currently exists that provides sufficiently effective selective capture so as to enable high capacity sorption of Mo-99-containing molybdenum, while simultaneously providing the selective separation of the Tc-99m decay product from the molybdenum in those cases where the Mo-99 has low specific activity. What is needed therefore is a device and methodology for selective generation of daughter radioisotope products from parent materials that has sufficient specificity so as to allow for lower specific activity parent materials to be utilized. The present invention is a significant advancement in this regard.